1. Field
The invention relates to compositions and methods for coating zirconium alloy cladding to enhance corrosion resistance and water resistance under nuclear reactor accident conditions and during normal operation.
2. Description of Related Art
In a typical commercial nuclear water reactor, such as a pressurized water reactor (PWR), heavy water reactor (e.g., a CANDU) or a boiling water reactor (BWR), the reactor core includes a large number of fuel assemblies, each of which is composed of a plurality, e.g., bundles or assemblies, of elongated fuel elements or fuel rods. Fuel assemblies vary in size and design depending on the desired size of the reactor and the core.
The fuel rods each contain nuclear fuel fissile material, such as, at least one of uranium dioxide (UO2), plutonium dioxide (PuO2), thorium dioxide (ThO2), uranium nitride (UN) and uranium silicide (U3Si2), and mixtures thereof. At least a portion of the fuel rods can also include neutron absorbing material, such as, boron or boron compounds, gadolinium or gadolinium compounds, erbium or erbium compounds and the like. The neutron absorbing material may be present on or in pellets in the form of a stack of nuclear fuel pellets. Annular or particle forms of fuel also can be used.
The fuel is encased in a sealed tube, commonly referred to as fuel cladding. Each of the fuel rods has cladding that acts as containment to hold the fissile material. The fuel rods are grouped together in an array which is organized to provide a neutron flux in the reactor core sufficient to support a high rate of nuclear fission and thus, the release of a large amount of energy in the form of heat. The cladding maintains the fuel in a position for which controlled fission can proceed and generate heat. A coolant, such as water, is pumped through the reactor core to extract the heat generated in the reactor core for the production of useful work such as electricity. The cladding then transfers the heat from the fuel to pressurized water that circulates around the primary loop of the reactor coolant system. The heated water in the primary loop is used to boil water in a steam generator and the steam is then expanded in a turbine that powers an electrical generator. Alternatively, the water circulating through the reactor may be allowed to boil to generate steam directly, which is then expanded in a turbine.
In a typical commercial nuclear reactor, the fuel assemblies in the core each have top and bottom nozzles. A plurality of elongated transversely spaced guide thimbles extends longitudinally between the nozzles. The plurality of elongated fuel elements or rods which compose the fuel assemblies are transversely spaced apart from one another and the guide thimbles. A plurality of transverse support grids are axially spaced along and attached to the guide thimbles. The grids are used to precisely maintain the spacing and support between the fuel rods in the reactor core, provide lateral support for the fuel rods, and induce mixing of the coolant.
FIG. 1 shows an exemplary reactor pressure vessel 10 and nuclear core 14. The nuclear core 14 includes a plurality of parallel, vertical, co-extending fuel assemblies 22. For purpose of this description, the other vessel internal structures can be divided into lower internals 24 and upper internals 26. In conventional designs, the lower internals' function is to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in FIG. 1), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 1, coolant enters the reactor pressure vessel 10 through one or more inlet nozzles 30, flows down through an annulus between the vessel and the core barrel 32, is turned 180° in a lower plenum 34, passes upwardly through a lower support plate 37 and a lower core plate 36 upon which the fuel assemblies 22 are seated and through and about the assemblies. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, a lower core support plate having the same elevation as 37. The coolant flow through the core and surrounding area 38 is typically large, on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting the nuclear core 14 flows along the underside of the upper core plate 40 and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially outward to one or more outlet nozzles 44.
One of the exemplary fuel assemblies 22 of FIG. 1 is shown in more detail in FIG. 2. As shown in FIG. 2, each of the fuel assemblies 22 includes radially-extending flukes or arms 52 and fuel rods 66 grouped in an array thereof. The fuel rods 66 are held in spaced relationship with one another by the grids 64 spaced along the fuel assembly length. At its lower end, a bottom nozzle 58 supports each of the fuel assemblies 22 on the lower core plate 36. At its upper end, each of the fuel assemblies 22 includes a top nozzle 62. An instrumentation tube 68 is located in the center and extends between and is mounted to the bottom and top nozzles 58 and 62, respectively. Each of the fuel rods 66 includes a plurality of nuclear fuel pellets 70 and is closed at its opposite ends by upper and lower end plugs 72 and 74, respectively. The pellets 70 are maintained in a stack by a plenum spring 76 disposed between the upper end plugs 72 and the top of the pellet stack. The fuel pellets 70, composed of fissile material, are responsible for creating the reactive power of the reactor.
One of the exemplary fuel rods 66 of FIG. 2 is shown in more detail in FIG. 3. As shown in FIG. 3, each of the fuel rods 66 includes a stack of the plurality of nuclear fuel pellets 70, the upper and lower end plugs 72 and 74, respectively, and the spring 76 which serves as a hold-down device to maintain the stacked configuration of the pellets 70. In addition, FIG. 3 shows the fuel rod cladding 2 which surrounds the pellets 70 to function as a barrier to prevent the fission by-products from entering the coolant and further contaminating the reactor system. The cladding 2 is typically in the shape of an elongated tube having a cavity formed therein and two opposing open ends. The upper and lower end plugs 72 and 74, respectively, provide a seal and prevent reactor coolant that is circulating in the core from entering the cavity of the fuel rod cladding 2. The thickness of the tube wall can vary. In certain embodiments, the tube wall thickness is from about 100 to about 1000 microns or from about 200 to 400 microns. The cladding may be composed of a zirconium (Zr)-based alloy. The cladding may include Zr and as much as about two percent by weight of other metals, such as niobium (Nb), tin (Sn), iron (Fe), chromium (Cr) and combinations thereof.
It is known in the art that there are various concerns relating to nuclear fuel rod cladding, including embrittlement of the cladding material during normal plant operation, which can lower the safety margin and potentially lead to failure under accident conditions, rapid corrosion of the Zr alloy tube at elevated temperatures associated with an accident scenario. In the event of an accident such as a Loss of Coolant Accident, temperatures inside the reactor core can exceed 1200° C. At very high temperatures, Zr rapidly oxidizes in the presence of steam which causes degradation of the fuel rods and production of large amounts of hydrogen which can lead to chemical explosions.
Surface-modification of the cladding is generally considered to improve corrosion resistance. Applying an oxidation-resistant coating to the outside surface of the cladding can at least reduce water corrosion and wear during normal plant operation, and potentially avoid potential negative consequences associated with Zr oxidation and fuel rod degradation in an accident scenario.
The coating can be applied on the fuel cladding using conventional coating methods, such as, but not limited to cold spraying and thermal spraying.
Cold spraying techniques generally include powder particles (e.g., the coating mixture including the master alloy, chemical activator and inert filler), typically, from about 10 to 50 μm, accelerated to very high velocities, typically, from 200 to 1000 m/s, by a compressed gas jet at temperatures well below their melting points. Upon impact with the substrate, the particles experience extreme and rapid deformation. This allows contact between the Zr alloy cladding surface under high local pressure, permitting bonding to occur and thick layers of deposited master alloy element(s) to build-up rapidly.
Thermal spraying techniques generally include thermal energy to melt or soften the coating mixture including the master alloy, chemical activator and inert filler under an inert atmosphere or a vacuum, causing the element(s) of the master alloy to adhere to the cladding surface and each other to form a coating. Thermal spray guns can be used to achieve the high velocity spraying.
It is an object of this invention to provide compositions and methods for coating a Zr alloy nuclear fuel cladding employing chromium, silicon, aluminum or mixtures thereof. The coating can be applied to an exterior surface of the fuel cladding, an interior surface of the fuel cladding, or both the exterior and the interior surfaces. The coating protects the exterior and/or interior surface(s) of the fuel cladding from one or more of oxidation, hydrogen uptake and wear failures. The coating is applied using a pack cementation method, which is traditionally employed in gas turbine and aircraft engine applications. Thicker coatings may be applied to Zr alloy tubes of greater thickness and then subjected to conventional cold work methods to reduce the thickness of both the coating and cladding. Thickness reduction may be achieved in several steps and heat treatment may be used between these steps to release residual stress in the material and to improve ductility.